The invention relates to a metallurgical process involving zirconium alloys, and more particularly to a process for treating zirconium alloys to immunize and improve resistance of such alloys to nodular corrosion when exposed to high pressure steam.
Nuclear reactors utilize water/steam as a coolant for the reactor as well as a source of energy to power steam turbines to thereby provide electrical energy. Nuclear reactors typically have their nuclear fissionable material contained in sealed cladding tubes, generally of a zirconium alloy, for isolation of the nuclear fuel from the water/steam. Zirconium and its alloys are widely used as nuclear fuel cladding since they advantageously possess low neutron absorption cross-sections, and at temperatures below about 398 C (the approximate core temperature of an operating nuclear reactor), are non-reactive and importantly possess high corrosion resistance relative to other metal alloys in the presence of de-mineralized water or steam. Two widely used zirconium alloys (xe2x80x9cZircaloysxe2x80x9d) are xe2x80x9cZircaloy-2xe2x80x9d and xe2x80x9cZircaloy-4xe2x80x9d, trade names of Westinghouse Electric Corporation for zirconium alloys of the above chemical compositions. Zircaloy-2, a Zrxe2x80x94Snxe2x80x94Nixe2x80x94Fexe2x80x94Cr alloy, is generally comprised (by weight) of approximately 1.2-1.7% tin, 0.13-0.20% iron, 0.06-0.15% chromium and 0.05-0.08% nickel. Zircaloy-4 has essentially no nickel, and about 0.2% iron, but is otherwise substantially similar to Zircaloy-2. Zircaloy-2 has enjoyed widespread use and continues to be used at present in nuclear reactors. Zircaloy-4 was developed as an improvement to Zircaloy-2 to reduce problems with hydriding, which causes Zircaloy-2 to become brittle when cooled to ambient temperatures (ie. when the reactor is shut down) after absorbing hydrogen at higher temperatures.
Zirconium alloys are among the best corrosion resistant materials when exposed to steam at reactor operating temperatures (less than 398 C, typically 290 C) in the absence of radiation from nuclear fission reactors. The corrosion rate in absence of neutron bombardment is very low and the corrosion product is a uniform, black ZrO2 oxide film/layer which forms on exterior surfaces of Zircaloy exposed to high temperature steam (uniform corrosion). The black oxide layer of ZrO2 usually contains a small (non-stoichiometric) excess of zirconium, and as such, it contains excess electrons giving it a black or gray color. It is also highly adherent to zirconium or Zircaloy surfaces exposed to steam.
Despite such relatively high corrosion resistance, when Zircaloys are used as cladding and exposed to high neutron flux in nuclear reactors, corrosion rates are generally increased, and cladding corrosion does become a potential problem in Pressurized Water Reactors (PWR""s) and particularly Boiling Water Reactors (BWR""s), where corrosion occurs in two formats, namely increased uniform corrosion as mentioned above, and alternatively, a second form, namely, nodular corrosion. Nodular corrosion is a highly undesirable, white, stoichiometric ZrO2 oxide layer (xe2x80x9cbloomxe2x80x9d) which forms on the surface of the cladding. It tends to form as small patches (xe2x80x9cnodulesxe2x80x9d or xe2x80x9cpustulesxe2x80x9d) on the surface of Zircaloys. Today, it is increasingly common to operate nuclear reactors at high xe2x80x9cburn-upxe2x80x9d (ie. to nearly complete consumption of the nuclear fuel). Under these conditions, the cladding is exposed to neutron flux for longer periods, which generally tends to increase the severity of nodular corrosion. Such increased nodular corrosion not only shortens the service life of the tube cladding (since when concentrated nodular corrosion acts in conjunction with certain contaminantsxe2x80x94such as copper ionsxe2x80x94localized spalling and ultimately penetration of the cladding can occur), but also produces a detrimental effect on the efficient operation of the reactor. In particular, the white ZrO2, being less adherent than black ZrO2, is prone to spalling or flaking away from the tube and entering into the reactor water, with detrimental effects. On the other hand, if the white nodular corrosion product does not spall away but remains on the tubing, a decrease in rapidity of heat transfer through the Zircaloy tube into the water cooling medium occurs when the less-dense white ZrO2 oxide layer covers an increasingly large portion of the Zircaloy tube exterior surface, and the reactor becomes less thermally efficient. Thus, nodular corrosion can become a significant problem for Zircaloy cladding in situations where Zircaloy tube cladding is left in the nuclear reactor for longer periods in conditions of high xe2x80x9cburn-upxe2x80x9d.
Zircaloys used in cladding for nuclear fuel rods are generally subject during their manufacture to a variety of heat treatments and anneals during the formation of the tubular cladding. It is known that the various heat treatments and quenching procedures used in forming a Zircaloy billet, and the various anneals and cold-working thereafter to form the Zircaloy tube cladding, all have an effect on the particular Zircaloy tubing""s ability to resist nodular corrosion, with some Zircaloys able to withstand nodular corrosion better than others despite both being of identical chemical composition. For example, fine grained equiaxed xcex1 Zircaloy-2, heated to 1010 C and slow-cooled at a rate of 18 C/hr. to 600 C and thereafter quenched, exhibits a high susceptibility to nodular corrosion under the standard steam test (510 C, 1500 psig, 24 hr.). Paradoxically, the same material, if simply quenched from 1010 C, or if heated to only 950 C and cooled at the same rate of 18 C/hr. to 600 C and thereafter quenched, exhibits high resistance to corrosion under the same standard steam test.
The actual physical changes in the structural properties of zirconium alloys during manufacturing processes of nuclear fuel tubing made therefrom were little understood, and it was therefore, prior to this invention, difficult to conceive of the best ways to immunize such fuel tubing to nodular corrosion. U.S. application Ser. No. 09/050,214 by the same inventor, filed Mar. 30, 1998 entitled xe2x80x9cMethod for Determining Corrosion Susceptibility of Nuclear Fuel Cladding to Nodular Corrosionxe2x80x9d, the subject matter of which is herein incorporated by reference, discloses that xcex1 Zircaloy-2 with very small precipitates, formed by having been heated to 1010 C and quenched, exhibits high resistance to nodular corrosion. Unfortunately, some research has suggested that small precipitates in the Zircaloy metal matrix can increase the danger of crack propagation in the cladding axial direction [see for example, U.S. patent application Ser. No. 08/052,793 entitled xe2x80x9cZircaloy Tubing Having High Resistance to Crack Propagationxe2x80x9d (now U.S. Pat. No. 5,519,748), and U.S. patent application Ser. No. 08/052,791 entitled xe2x80x9cMethod of Fabricating Zircaloy Tubing Having High Resistance to Crack Propagationxe2x80x9d (now U.S. Pat. No. 5,437,747), both assigned to the assignee hereof]. Thus, while zirconium alloy tubing possessing excellent resistance to nodular corrosion may be manufactured, it is frequently necessary to add further annealing heat treatments to achieve other further objectives, such as to reduce the incidence of axial splitting of a Zircaloy-2 tubing. Unfortunately, up until the present invention and the understanding of the concept of critical temperature Tc disclosed in U.S. application Ser. No. 09/050,214, filed Mar. 30, 1998, entitled xe2x80x9cMethod for Determining Corrosion Susceptibility of Nuclear Fuel Cladding to Nodular Corrosionxe2x80x9d, such other anneal processes often had detrimental effects on the ability of such zirconium alloy tubing to withstand nodular corrosion. In fact, until the present invention, it was little understood why some annealing processes actually have the effect of sensitizing the tubing to nodular corrosion. It was thus unknown, prior to this invention, how to reliably retain the benefits of a zirconium alloy possessing high resistance to nodular corrosion when further subjecting such tubing to a further anneal for the purposes of increasing such tubing""s resistance to axial splitting. It was further unknown, prior to this invention, how to reliably apply an anneal to a zirconium alloy (which may initially prior to such anneal be susceptible to nodular corrosion) so as to completely immunize against nodular corrosion.
The present invention involves an understanding of known metallurgical structures of Zircaloys, namely xcex1 and xcex2 crystal lattice structures which are present, either individually or in combination, over certain temperature ranges during annealing of Zircaloy.
The specific xcex1 or xcex2 metallurgical structures which are present over various temperature ranges are commonly depicted in time-temperature plots referred to as isothermal transformation diagrams (xe2x80x9cTTTxe2x80x9d diagrams) specific to each Zircaloy. In particular, such diagrams refer to a xe2x80x9cxcex2 phasexe2x80x9d, which for zirconium alloys is a body-centered cubic crystal lattice structure of crystalline zirconium, which exists at temperatures above about 825 C, and exclusively exists at temperatures above about 985 C. Such diagrams further refer to the xcex1-matrix phase, or xe2x80x9cxcex1 phasexe2x80x9d. The xcex1-matrix phase of a Zircaloy-2 comprises a close-packed hexagonal lattice structure, which exists exclusively at temperatures less than approximately 825 C. Both the xcex1 and xcex2 phases of Zircaloys may simultaneously each exist at temperatures in the range of about 825-985 C.
Precipitates (herein referred to collectively by the Greek letter "khgr") also exist within Zircaloys, and are generally particles within the alloy containing higher concentrations of the alloying elements Fe and Cr or Ni. These alloying elements, which exist in solution at low concentrations in the xcex1 and xcex2 matrices, generally start to precipitate out of solution and form precipitates below temperatures of about 855 C.
Precipitates found in Zircaloys are represented by chemical formulas such as Zr (Fe, Cr)2 and Zr2 (Fe, Ni).
For a given temperature less than about 855 C [ie. for a given temperature less than the temperature of the (xcex1+xcex2)/((xcex1+xcex2precipitate) transus on the TTT diagram], at equilibrium the concentration of alloying elements Fe, Cr and Ni (ie. solute) in the xcex1-matrix will be no higher than their solubility limit within such xcex1 matrix at the given temperature. Lowering the temperature causes such solute to precipitate out of the xcex1-matrix into precipitates and/or to migrate to remaining xcex1 phase, if xcex1 phase exists at such temperature.
It is postulated that Zircaloys derive their immunity to nodular corrosion from solute present in the xcex1-matrix, the xcex1-matrix being the metallurgical structure present at the temperature at which nuclear reactors operate (ie. in the 200 C-390 C range). Zircaloys which are rapidly cooled from a relatively high temperature (eg. 950 C) [when large amounts of solute may be present in such xcex1-matrix and where such solutes remain trapped in such xcex1-matrix in a supersaturated condition when subsequently rapidly cooled] have high resistance to nodular corrosion, which lends support to such postulate. Likewise supporting such postulate is evidence that slowly cooling a Zircaloy from temperatures commencing at 950 C, whereby the concentration of solute in such xcex1-matrix is thereby given the chance to leave such matrix during the cooling period by, for example, precipitating into precipitates, produces a Zircaloy having an xcex1-matrix phase at temperatures of 200 C-390 C which is highly sensitive to nodular corrosion.
It is now believed that there exists a critical concentration Cc of solute within such xcex1-matrix whereby xcex1-matrices having concentrations of solute therein above such critical concentration Cc will possess an immunity to nodular corrosion, and xcex1-matrices having solute concentrations below such critical concentration Cc will exhibit high susceptibility to nodular corrosion when exposed to steam. Experimental tests conducted by the inventor have allowed the inventor to conclude the critical concentration Cc of solute is reached, when at equilibrium, at a specific critical temperature Tc which, for Zircaloys, exists in the region below the (xcex1+xcex2)/(xcex1+xcex2+"khgr") transus (ie. below about 855 C) but above the (xcex1+xcex2+"khgr")/(xcex1+"khgr") transus (ie. above about 825 C), namely in the (xcex1+xcex2+"khgr") region on the TTT diagram. In particular, experimental results conducted with Zircaloy-2 indicate this critical temperature Tc to be in the range of 837-841 C, and likely about 840 C. At such temperature the xcex1-matrix containing solute of a concentration Cc exists in equilibrium with solute-saturated xcex2-phase.
Axial splitting of Zircaloy tubing is a further recognized problem. Axial splitting on the surface of such tubing leads to localized stress concentrations and increased corrosion in such cracks, leading to splitting of tubing and thereby contamination of the reactor coolant by the radioactive fuel. It is recognized that an anneal applied to the surface of zirconium alloy tubing which is exposed to water/steam in a nuclear reactor, at a temperature below about 855 C, [namely at a temperature on the TTT diagram in the region where precipitates form, namely in the xcex1+xcex2, xcex1+xcex2+"khgr" and xcex1+"khgr" regions] but above the critical temperature, in addition to immunizing against nodular corrosion will cause precipitates formed within such alloy below such temperature to coarsen, namely grow larger in size, and such has the beneficial result of reducing instances of axial splitting of such tubing on the surface of such tubing (hereinafter referred to as a xe2x80x9ccoarsening annealxe2x80x9d).
Using the concept of Tc, it is now realized that annealing a zirconium alloy at a temperature above the critical temperature Tc, but below approximately 855 C where precipitates form, namely below the xcex1+xcex2+precipitate)/(xcex1+xcex2) transus temperature for the particular zirconium alloy, will, when subsequently rapidly cooled, results in an alloy possessing both coarsened precipitates and at the same time a resistance/immunity to nodular corrosion. Likewise, it is now recognized, using the concept of Tc, that annealing a zirconium alloy at a temperature above the critical temperature Tc but below the (xcex1+xcex2+precipitate)/(xcex1+xcex2) transus temperature for the particular zirconium alloy will even cause a zirconium alloy not originally possessing immunity to nodular corrosion to then possess an immunity to nodular corrosion by coarsening of any and all intermetallic grain structures.
Accordingly, in one of its broad aspects the present invention teaches a method of immunizing and/or increasing resistance of a surface of zirconium alloy tubing exposed to high pressure steam to nodular corrosion.
In another of its broad aspects the present invention comprises a method of immunizing and/or increasing resistance of a surface of zirconium alloy tubing to nodular corrosion, while at the same time coarsening precipitates formed during the anneal process.
More particularly, in one of its broad aspects the present invention teaches a method of immunizing and/or increasing resistance of a surface of zirconium alloy tubing to nodular corrosion, comprising:
(i) heating said surface of said tubing to a temperature within a temperature range bounded at its lower limit by a temperature Tc, namely the temperature which at equilibrium conditions sufficient solute would exist in xcex1-matrices of said zirconium alloy to resist nodular corrosion, and bounded at its upper limit by a temperature of the (xcex1+xcex2+precipitate)/(xcex1+xcex2) transus inherent for the particular zirconium alloy;
ii) maintaining the temperature of said surface within said temperature range for a selected period of time; and
iii) subsequently causing said temperature of said surface to be reduced at a rate sufficiently rapid to prevent substantial loss of solute concentration from said xcex1-matrices.
In another of its broad aspects, the present invention thus teaches a method of increasing resistance of a surface of zirconium alloy tubing to nodular corrosion, comprising:
(i) heating such surface of said tubing to a temperature within a temperature range bounded at its lowest limit by a temperature Tc, namely the temperature at which at equilibrium conditions sufficient solute would exist in xcex1-matrices of said zirconium alloy to resist nodular corrosion, and bounded at its upper limit by a temperature of the (xcex1+xcex2+precipitate)/(xcex1+xcex2) transus on the isothermal transformation temperature diagram for the particular zirconium alloy;
(ii) maintaining said surface within said temperature range for a time period sufficient to allow an increase in the average size of a portion of intermetallic particles formed within said alloy at said surface; and
(iii) subsequently causing said temperature of said surface to be reduced at a rate sufficiently rapid to prevent substantial loss of solute concentration in said xcex1-matrices.
In still a further aspect of the present invention, a zirconium alloy tubing is taught, said tubing comprising:
(i) a surface which has been heated to a temperature within a temperature range bounded at its lower limit by a temperature Tc, namely the temperature at which at equilibrium conditions sufficient solute would exist in xcex1-matrices of said zirconium alloy to resist nodular corrosion, and bounded at its upper limit by a temperature of the (xcex1+xcex2+precipitate)/(xcex1+xcex2) transus on the isothermal transformation temperature diagram for the particular zirconium alloy; and
(ii) said surface having been maintained within said temperature range for a time period sufficient to allow an increase in the size of intermetallic particles formed within said surface, and subsequently thereafter cooled.
In each of the above three aspects of the invention, where the zirconium alloy is Zircaloy-2 comprised of zirconium and the following metals in approximate weight percentages, namely 1.2-1.7% tin, 0.13-0.20% iron, 0.06-0.15% chromium, and 0.05-0.08% nickel, the corresponding critical temperature Tc is in the range of about 837-841 C, preferably approximately 840 C, and the temperature of the (xcex1+xcex2+precipitate)/(xcex1+xcex2) transus is approximately 855 C.
The holding time pursuant to which such surface is held within such temperature range may vary within the range of anywhere from greater than 2 seconds to over 20 hours. For example, the holding time may be greater than 20 seconds, greater than 30 seconds, greater than 1 minute, and up to 20 hours or more. Preferably the holding time is in the range of approximately 30 minutes to two hours, to give sufficient time to allow intermetallic particles to increase in size and to immunize against nodular corrosion. Typically, the step of reducing the temperature after exposure to such temperature comprises cooling the surface by quenching. During quenching, the temperature may be reduced at a rate of at least 1xc2x0 C./sec., typically at least 3xc2x0 C./sec, for example.
The special protective anneal of the present invention, due to being carried out at a temperature range not exceeding the temperature of the (xcex1+xcex2+precipitate)/xcex1+xcex2) transus on the isothermal transformation temperature diagram, is able to increase resistance to nodular corrosion with the original precipitates present. Annealing at a higher temperature anneal above not only Tc but also above the (xcex1+xcex2+precipitate)/(xcex1+xcex2) transus would otherwise re-dissolve all precipitates as no precipitates may exist above such transus. Upon cooling such would leave a distribution of very small new precipitates and not produce the more desirable larger size precipitates as in the case of the present invention, which, it is believed, reduces the susceptibility of the Zircaloy-2 tubing to axial splitting.